Assessment of Occupational Exposure Due to Intakes of Radionuclides | IAEA

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IAEA SAFETY STANDARDS

SERIES

Assessment of

Occupational Exposure Due to Intakes of

Radionuclides

JOINTLY SPONSORED BY THE

INTERNATIONAL ATOMIC ENERGY AGENCY AND THE INTERNATIONAL LABOUR OFFICE

SAFETY GUIDE

No. RS-G-1.2

INTERNATIONAL

ATOMIC ENERGY AGENCY

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IAEA SAFETY RELATED PUBLICATIONS

IAEA SAFETY STANDARDS

Under the terms of Article III of its Statute, the IAEA is authorized to establish standards of safety for protection against ionizing radiation and to provide for the application of these standards to peaceful nuclear activities.

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ASSESSMENT OF OCCUPATIONAL EXPOSURE

DUE TO INTAKES OF

RADIONUCLIDES

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The following States are Members of the International Atomic Energy Agency:

AFGHANISTAN ALBANIA ALGERIA ARGENTINA ARMENIA AUSTRALIA AUSTRIA BANGLADESH BELARUS BELGIUM BENIN BOLIVIA BOSNIA AND

HERZEGOVINA BRAZIL BULGARIA BURKINA FASO CAMBODIA CAMEROON CANADA CHILE CHINA COLOMBIA COSTA RICA COTE D’IVOIRE CROATIA CUBA CYPRUS

CZECH REPUBLIC DEMOCRATIC REPUBLIC

OF THE CONGO DENMARK

DOMINICAN REPUBLIC ECUADOR

EGYPT EL SALVADOR ESTONIA ETHIOPIA FINLAND FRANCE GABON GEORGIA GERMANY GHANA GREECE

GUATEMALA HAITI HOLY SEE HUNGARY ICELAND INDIA INDONESIA

IRAN, ISLAMIC REPUBLIC OF IRAQ

IRELAND ISRAEL ITALY JAMAICA JAPAN JORDAN KAZAKHSTAN KENYA

KOREA, REPUBLIC OF KUWAIT

LATVIA LEBANON LIBERIA

LIBYAN ARAB JAMAHIRIYA LIECHTENSTEIN

LITHUANIA LUXEMBOURG MADAGASCAR MALAYSIA MALI MALTA

MARSHALL ISLANDS MAURITIUS

MEXICO MONACO MONGOLIA MOROCCO MYANMAR NAMIBIA NETHERLANDS NEW ZEALAND NICARAGUA NIGER NIGERIA NORWAY PAKISTAN

PANAMA PARAGUAY PERU PHILIPPINES POLAND PORTUGAL QATAR

REPUBLIC OF MOLDOVA ROMANIA

RUSSIAN FEDERATION SAUDI ARABIA SENEGAL SIERRA LEONE SINGAPORE SLOVAKIA SLOVENIA SOUTH AFRICA SPAIN

SRI LANKA SUDAN SWEDEN SWITZERLAND

SYRIAN ARAB REPUBLIC THAILAND

THE FORMER YUGOSLAV REPUBLIC OF MACEDONIA TUNISIA

TURKEY UGANDA UKRAINE

UNITED ARAB EMIRATES UNITED KINGDOM OF

GREAT BRITAIN AND NORTHERN IRELAND UNITED REPUBLIC

OF TANZANIA

UNITED STATES OF AMERICA URUGUAY

UZBEKISTAN VENEZUELA VIET NAM YEMEN YUGOSLAVIA ZAMBIA ZIMBABWE

The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957. The Headquarters of the Agency are situated in Vienna. Its principal objective is “to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world’’.

Permission to reproduce or translate the information contained in this publication may be obtained by writing to the International Atomic Energy Agency, Wagramer Strasse 5, P.O. Box 100, A-1400 Vienna, Austria.

Printed by the IAEA in Austria October 1999 STI/PUB/1077

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ASSESSMENT OF

OCCUPATIONAL EXPOSURE DUE TO INTAKES OF

RADIONUCLIDES

SAFETY GUIDE

JOINTLY SPONSORED BY THE

INTERNATIONAL ATOMIC ENERGY AGENCY AND THE INTERNATIONAL LABOUR OFFICE

SAFETY STANDARDS SERIES No. RS-G-1.2

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1999

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VIC Library Cataloguing in Publication Data

Assessment of occupational exposure due to intakes of radionuclides : safety guide. — Vienna : International Atomic Energy Agency, 1999.

p. ; 24 cm. — (Safety standards series, ISSN 1020–525X ; no. RS-G-1.2) STI/PUB/1077

ISBN 92–0–101999–8

Includes bibliographical references.

1. Radiation workers. 2. Radiation dosimetry. I. International Atomic Energy Agency. II. Series.

VICL 99–00225

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FOREWORD by Mohamed ElBaradei

Director General

One of the statutory functions of the IAEA is to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes, and to provide for the application of these standards to its own operations as well as to assisted operations and, at the request of the parties, to operations under any bilateral or multilateral arrangement, or, at the request of a State, to any of that State’s activities in the field of nuclear energy.

The following advisory bodies oversee the development of safety standards: the Advisory Commission on Safety Standards (ACSS); the Nuclear Safety Standards Advisory Committee (NUSSAC); the Radiation Safety Standards Advisory Committee (RASSAC); the Transport Safety Standards Advisory Committee (TRANSSAC); and the Waste Safety Standards Advisory Committee (WASSAC).

Member States are widely represented on these committees.

In order to ensure the broadest international consensus, safety standards are also submitted to all Member States for comment before approval by the IAEA Board of Governors (for Safety Fundamentals and Safety Requirements) or, on behalf of the Director General, by the Publications Committee (for Safety Guides).

The IAEA’s safety standards are not legally binding on Member States but may be adopted by them, at their own discretion, for use in national regulations in respect of their own activities. The standards are binding on the IAEA in relation to its own operations and on States in relation to operations assisted by the IAEA. Any State wishing to enter into an agreement with the IAEA for its assistance in connection with the siting, design, construction, commissioning, operation or decommissioning of a nuclear facility or any other activities will be required to follow those parts of the safety standards that pertain to the activities to be covered by the agreement.

However, it should be recalled that the final decisions and legal responsibilities in any licensing procedures rest with the States.

Although the safety standards establish an essential basis for safety, the incorporation of more detailed requirements, in accordance with national practice, may also be necessary. Moreover, there will generally be special aspects that need to be assessed by experts on a case by case basis.

The physical protection of fissile and radioactive materials and of nuclear power plants as a whole is mentioned where appropriate but is not treated in detail;

obligations of States in this respect should be addressed on the basis of the relevant instruments and publications developed under the auspices of the IAEA.

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Non-radiological aspects of industrial safety and environmental protection are also not explicitly considered; it is recognized that States should fulfil their international undertakings and obligations in relation to these.

The requirements and recommendations set forth in the IAEA safety standards might not be fully satisfied by some facilities built to earlier standards. Decisions on the way in which the safety standards are applied to such facilities will be taken by individual States.

The attention of States is drawn to the fact that the safety standards of the IAEA, while not legally binding, are developed with the aim of ensuring that the peaceful uses of nuclear energy and of radioactive materials are undertaken in a manner that enables States to meet their obligations under generally accepted principles of international law and rules such as those relating to environmental protection. According to one such general principle, the territory of a State must not be used in such a way as to cause damage in another State. States thus have an obligation of diligence and standard of care.

Civil nuclear activities conducted within the jurisdiction of States are, as any other activities, subject to obligations to which States may subscribe under international conventions, in addition to generally accepted principles of international law.

States are expected to adopt within their national legal systems such legislation (including regulations) and other standards and measures as may be necessary to fulfil all of their international obligations effectively.

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PREFACE

Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy.

In 1996, the Agency published Safety Fundamentals on Radiation Protection and the Safety of Radiation Sources (IAEA Safety Series No. 120) and International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (IAEA Safety Series No. 115), both of which were jointly sponsored by the Food and Agriculture Organization of the United Nations, the IAEA, the International Labour Organisation, the OECD Nuclear Energy Agency, the Pan American Health Organization and the World Health Organization. These publications set out, respectively, the objectives and principles for radiation safety and the requirements to be met to apply the principles and to achieve the objectives.

The establishment of safety requirements and guidance on occupational radiation protection is a major component of the support for radiation safety provided by the Agency to its Member States. The objective of the Agency’s Occupational Protection Programme is to promote an internationally harmonized approach to the optimization of occupational radiation protection, through the development and application of guidelines for restricting radiation exposures and applying current radiation protection techniques in the workplace.

Guidance on meeting the requirements of the Basic Safety Standards for occupational protection is provided in three interrelated Safety Guides, one giving general guidance on the development of occupational radiation protection programmes and two giving more detailed guidance on the monitoring and assessment of workers’ exposure due to external radiation sources and from intakes of radionuclides, respectively. These Safety Guides together reflect the current internationally accepted principles and recommended practices in occupational radiation protection, with account taken of the major changes that have occurred over the past decade.

The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office.

The present Safety Guide addresses the assessment of exposure due to intakes of radionuclides in the workplace. Such intakes can occur via a number of pathways whenever unsealed sources are present, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to intakes depends critically upon knowledge of the biokinetics of the radionuclides, and the present Safety Guide reflects the major changes over the past decade in international practice in internal dose assessment.

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EDITORIAL NOTE

An appendix, when included, is considered to form an integral part of the standard and to have the same status as the main text. Annexes, footnotes and bibliographies, if included, are used to provide additional information or practical examples that might be helpful to the user.

The safety standards use the form ‘shall’ in making statements about requirements, responsibilities and obligations. Use of the form ‘should’ denotes recommendations of a desired option.

The English version of the text is the authoritative version.

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1. INTRODUCTION

BACKGROUND

1.1. Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved.

1.2. The IAEA Safety Fundamentals publication Radiation Protection and the Safety of Radiation Sources [1] presents the objectives, concepts and principles of radiation protection and safety. Requirements designed to meet the objectives and apply the principles specified in the Safety Fundamentals, including requirements for the protection of workers exposed to sources of radiation, are established in the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (commonly referred to as the Basic Safety Standards or BSS), jointly sponsored by the Agency and five other international organizations [2].

1.3. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure.

Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials.

1.4. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5].

These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide.

OBJECTIVE

1.5. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the

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planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring.

SCOPE

1.6. This Safety Guide presents the main considerations for monitoring for internal exposures in both routine and accident situations, using direct and indirect methods.

It also introduces monitoring of levels of radionuclides in the working environment as a basis for assessing intakes. The biokinetic and dosimetric models needed for more specific estimates of doses to individuals, to be used in the case of accidents or incidents, or when operations could result in doses approaching regulatory limits, are also presented.

1.7. This Safety Guide does not cover the medical exposure of patients or exposure of members of the public, nor does it give specific advice on monitoring of workers in mining and milling.

1.8. Technical details and advice on the assessment of internal contamination by direct methods has been published by the IAEA [7]. Practical advice on the use of indirect methods as well as interpretation of measurements in terms of the amount of radioactive material taken into the body and the associated radiation doses will be given in future IAEA publications.

STRUCTURE

1.9. The primary dosimetric and derived operational quantities used in radiation protection that relate to the assessment of doses from intakes of radionuclides are summarized in Section 2. In Section 3, the principles involved in the development of monitoring programmes and the need for individual or area monitoring are discussed.

The selection of individuals and the choice of either direct or indirect methods for assessing the extent of any internal contamination in routine and accident situations are also reviewed in Section 3. The methods that have been developed for directly assessing the body or organ/tissue content of radionuclides by external counting of photon emissions emanating from the body are reviewed in Section 4. The use of indirect methods for assessing either the body content of a radionuclide or for investigat- ing whether an intake has occurred from biological or physical samples is considered in Section 5. Models for describing the behaviour of radionuclides in the body are summarized in Section 6. Their application to calculate levels of radionuclides in the

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body and hence radiation doses from measurements made by either direct or indirect methods is illustrated in Section 7. Requirements for record keeping, for measurements both on individuals and from area monitoring, are considered in Section 8.

Finally, guidance on quality assurance procedures is given in Section 9.

1.10. Two appendices and an annex provide additional information. Appendix I provides suggested criteria to indicate whether individual monitoring is necessary.

Appendix II defines procedures for calculating detection limits for measurement methods. The Annex provides, for ease of reference, some basic data relevant to the assessment of occupational exposure due to intakes of radionuclides, namely tissue weighting factors and dose coefficients (committed doses per unit intake) and derived air concentrations (DACs) for selected chemical forms of some common radionuclides.

2. DOSIMETRIC QUANTITIES

2.1. The quantities adopted in the BSS to express the doses received from intakes of radionuclides for radiological protection purposes are the effective dose E and the equivalent dose H_Tin tissue or organ T. These quantities are briefly discussed in the related Safety Guide [3] and are formally defined in the BSS [2]. The quantity of primary interest for internal dose assessment is the intake, which is defined here as the activity of a radionuclide taken into the body¹. The dose coefficient (committed effective dose per unit intake) for radionuclide j — by ingestion, e(g)_j,ing, or by inhalation,e(g)_j,inh, as appropriate — is used to determine the committed effective dose from an estimated intake. For occupational exposure, all exposed persons are adults and therefore the period of time over which the committed effective dose is assessed is 50 years, irrespective of the age at intake.

2.2. Internal doses cannot be measured directly; they can only be inferred from measured quantities such as body activity content, excretion rates or airborne concentrations of radioactive material. Section 7 provides an illustration of the assessment of doses from such measurements.

1 In the BSS, intake is defined as “the process of taking radionuclides into the body by inhalation or ingestion or through the skin”. In this Safety Guide, the term intake is used both in this descriptive sense and in the more specific sense described in the text.

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2.3. In situations of exposure due to a single radionuclide by inhalation or ingestion, with no external exposure, the limit on intake I_j,Lcorresponding to the relevant limit Lon effective dose is given by:

where e(g)_jis the relevant value of committed effective dose per unit intake. When there is exposure due to a range of radionuclides and/or external exposure, the total effective dose will need to be calculated. Requirements for and guidance on dose assessment in these circumstances are given in the BSS [2] and in Ref. [3].

2.4. Values of the committed effective dose per unit intake by ingestion and by inhalation for occupational exposure are given in Table II–III of the BSS [2] (except for radon progeny and thoron progeny). Values for selected radionuclides are reproduced in Table A–I of the Annex to the present Safety Guide.

2.5. The limits on intake and exposure for radon progeny and thoron progeny are given in Table II–I of the BSS [2] and are summarized in the companion Safety Guide [3].

2.6. The fraction of an intake that remains in the body (for direct methods) or that is being excreted from the body (for indirect methods) at time tafter an intake may be designated m(t) [8, 9]. This fraction depends on the radionuclide, its chemical and physical form and the route of intake, as well as t. To estimate the intake for dose assessment, the measured body content or excretion rate must be divided by the appropriate value of m(t) (see Section 7). The committed dose can be seriously underestimated if the dose coefficient e(g)_jis applied directly to the measured body content rather than to the inferred intake.

2.7. The potential for inhalation of radionuclides should be assessed when necessary by measuring activity levels in air samples. The derived air concentration (DAC, expressed in Bq/m³) is defined as that concentration of airborne activity which would result in the intake of I_j,inh,Lby a worker exposed continuously for one year (taken to be 2000 working hours). For a standard breathing rate of 1.2 m³/h, the DAC would thus be given by:

DAC=

× I 2000 1.2

j,inh,L

I L

j,L e(g)

j

=

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2.8. For example, for inhalation by a worker of ¹³⁷Cs as an aerosol with an AMAD of 5 µm,e(g)_j,inhis 6.7 ×10^–9Sv/Bq. If Lis assumed to be the occupational dose limit of 20 mSv/a (0.02 Sv/a) [3], then:

and

In practice, the DAC would be rounded to 1 ×10³Bq/m³. Table A–2 of the Annex gives example values of DACs.

2.9. The measured airborne activity concentration, expressed as a fraction of the DAC, may be multiplied by the exposure time in hours to obtain an estimate of intake expressed in units of DAC⋅h. By definition, 2000 DAC·h corresponds to an intake of I_j,inh,L.

3. MONITORING PROGRAMME

GENERAL OBJECTIVE

3.1. The general objective of operational monitoring programmes is the assessment of workplace conditions and individual exposures. The assessment of doses to workers routinely or potentially exposed to radiation through intakes of radioactive material constitutes an integral part of any radiation protection programme and helps to ensure acceptably safe and satisfactory radiological conditions in the workplace.

3.2. Measures to meet the general requirements for the radiation protection of workers are described in the related Safety Guide on the application of occupational radiation protection principles [3]. The specific aspects of monitoring that relate to exposure due to intakes of radioactive material are described below.

DAC= × Bq m

× = ×

3 10 1 3 10

6

3 3

2000 1.2 . /

I_j,inh,L =

× ₋ = × 0 02

6 7 10. ₉ 3 10⁶

. Bq

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INDIVIDUAL DOSE ASSESSMENT

3.3. Typical methods of individual monitoring for intakes are whole body counting, organ counting (such as thyroid or lung monitoring) and analysis of samples of excreta. Sampling of the breathing zone with personal air samplers is also used.

3.4. In many circumstances involving exposure due to radionuclides, workplace monitoring will be needed. Monitoring procedures may be introduced to demonstrate satisfactory working conditions or in cases where individual monitoring is unable to provide adequate protection of the worker. Such workplace monitoring may also be appropriate when levels of contamination are low, for example in a research laboratory using small quantities of radioactive tracers.

3.5. Monitoring for the estimation of doses from intakes of radionuclides may include one or more of the following techniques:

(a) Sequential measurements of radionuclides in the whole body or in specific organs;

(b) Measurements of radionuclides in biological samples such as excreta or breath;

(c) Measurement of radionuclides in physical samples such as filters from personal or fixed air samplers, or surface smears.

Measurements can be used to calculate the intake of a radionuclide, which, when multiplied by the appropriate dose coefficient, leads to an estimate of committed effective dose. Dose coefficients for a wide range of radionuclides are given in the BSS [2], and those for selected radionuclides are reproduced in Table A–1 of the Annex. In some circumstances, the results of direct measurement may be used to calculate dose rates to the whole body or to individual organs.

Need for monitoring

3.6. The designation of areas as controlled areas or supervised areas and the need for individual monitoring will be determined from a knowledge of conditions in the workplace and the potential for worker exposure. In general, the decision to enrol a worker in an internal exposure monitoring programme should be based on the likelihood that the individual could receive an intake of radioactive material exceeding a predetermined level. Guidance on the designation of controlled and supervised areas is given in the related Safety Guide [3]. If operational procedures are set up to prevent or reduce the possibility of intake, a controlled area will, in general, need to be established [3].

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3.7. The need or otherwise for individual or area monitoring for internal exposure will depend on the amount of radioactive material present and the radionuclide(s) involved, the physical and chemical form of the radioactive material, the type of containment used, the operations performed and the general working conditions. For example, workers handling sealed sources, or unsealed sources in reliable containment, may need to be monitored for external exposure, but not necessarily for internal exposure. Conversely, workers handling radionuclides such as tritium,¹²⁵I or ²³⁹Pu may need monitoring for internal exposure, but not for external exposure.

3.8. It can be difficult to determine whether monitoring a worker for intakes of radioactive material is necessary. Such monitoring should be used routinely only for workers who are employed in areas that are designated as controlled areas specifically in relation to the control of contamination and in which there are grounds for expect- ing significant intakes. If experience has shown that it is unlikely that committed effective doses from annual intakes of radionuclides from occupational exposure would exceed 1 mSv, then individual monitoring may be unnecessary, but workplace monitoring should be undertaken.

3.9. Examples of situations in which experience has shown that it is necessary to give consideration to routine individual monitoring for internal exposure include the following:

(a) Handling of large quantities of gaseous or volatile materials, for example tritium and its compounds in large scale production processes, in heavy water reactors and in luminizing;

(b) Processing of plutonium and other transuranic elements;

(c) Mining, milling and processing of thorium ores, and the use of thorium and its compounds (which can lead to internal exposure due both to radioactive dusts and to thoron (²²⁰Rn) and its progeny);

(d) Mining, milling and refining of high grade uranium ores;

(e) Processing of natural and slightly enriched uranium, and reactor fuel fabrication;

(f) Bulk production of radioisotopes;

(g) Working in mines and other workplaces where radon levels exceed a specified action level;

(h) Handling of large quantities of radiopharmaceuticals, such as ¹³¹I for therapy;

(i) Maintenance of reactors, which can lead to exposure due to fission and activation products.

3.10. For some radionuclides, individual monitoring may not be feasible because of the radiation type(s) emitted and the detection sensitivity of monitoring methods, and reliance must be placed on workplace monitoring. Conversely, for some other

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radionuclides such as tritium, individual monitoring may be more sensitive than workplace monitoring.

3.11. For new operations, individual monitoring is likely to be needed and should be considered. As experience in the workplace is accumulated, the need for routine individual monitoring should be kept under review. Workplace monitoring may be found to be sufficient for radiological protection purposes.

3.12. Some guidance on and examples of criteria to determine whether individual monitoring is necessary are given in Appendix I.

Design of a routine monitoring programme

3.13. Routine internal exposure monitoring is that conducted on a fixed schedule for selected workers. Internal exposure monitoring has several limitations that should be considered in the design of an adequate monitoring programme.

3.14. Firstly, monitoring does not measure directly the committed effective dose to the individual. Biokinetic models are needed to relate the activity level in an excreta sample to that in the body at the time the sample was taken, to relate the body content at the time the sample was taken to the original intake, and to calculate the committed effective dose from the estimated intake.

3.15. Secondly, measurements may be subject to interference from other radionuclides present in the body, such as ⁴⁰K present naturally,¹³⁷Cs from global fallout, uranium naturally present in the diet or radiopharmaceuticals administered for diag- nostic or therapeutic purposes. It is therefore important to establish the body content of both naturally occurring and artificial radionuclides from previous intakes. This is particularly important when the non-occupational intakes are elevated, for example in mining areas having above average domestic exposure due to radon. All workers should undergo bioassay measurements before commencing work with radioactive materials, in order to determine a ‘background’ level.

3.16. Radiopharmaceuticals may interfere with bioassay measurements for some time after administration, depending on the properties of the agent administered and on the radionuclides present at the workplace. Workers should be requested to report any administration of radiopharmaceuticals to their supervisors, so that it can be determined whether or not adequate internal exposure monitoring can be performed.

3.17. Thirdly, the results of an individual monitoring programme for the estimation of chronic intakes might depend on the time at which the monitoring is performed.

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For certain radionuclides with a significant early clearance component of excretion, there may be a significant difference between measurements taken before and after the weekend. Such cases should be reviewed individually if chronic exposure is possible [8–10]. Additionally, for nuclides with long effective half-lives, the amount present in the body and the amount excreted depend on, and will increase with, the number of years for which the worker has been exposed. In general, the retained activity from previous years’ intakes should be taken to be part of the background for the current year.

3.18. Finally, the analytical methods used for individual monitoring sometimes do not have adequate sensitivity to detect the activity levels of interest (see Appendix II).

If individual monitoring is not feasible, then a system of workplace and personnel monitoring should be employed to determine, as far as possible, the amounts of radionuclides that may have been taken in by an individual. Fixed (static) air samplers or personal air samplers (PASs) may be used to determine the concentration of airborne radioactive material, which can be combined with standard or site specific assumptions about the physicochemical form of the material and the breathing rate and exposure time of the worker to estimate inhalation intakes. Similarly, surface monitoring may be used to indicate the potential for intake or the need for more detailed area monitoring, but the models for estimating intakes from surface contamination are particularly uncertain.

3.19. Exposure due to radon is of particular concern in underground mines, in buildings constructed with material containing significant levels of radium, in offices, factories and other premises with elevated levels of uranium in the ground, and in buildings where large amounts of groundwater are processed. In 1993, the ICRP issued recommendations on protection against ²²²Rn at home and at work [11]. Safety Series No. 26 [12] covers radiation protection in the mining and milling of radioactive ores.

Methods of measurement

3.20. Intakes of radionuclides can be determined by either direct or indirect measurement methods. Direct measurements of gamma or X ray photons (including bremsstrahlung) emitted from internally deposited radionuclides are frequently referred to as body activity measurements, whole body monitoring or whole body counting. Indirect measurements are measurements of activity in samples which may be either biological (e.g. excreta) or physical (e.g. air filters). Each type of measurement has advantages and disadvantages, and the selection of one rather than another is largely dependent on the nature of the radiation to be measured. Direct methods are useful only for those radionuclides which emit photons of sufficient energy, and in sufficient numbers, to escape from the body and be measured by an external detector.

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Many fission and activation products fall into this category. Incorporated radionuclides which do not emit energetic photons (e.g. ³H,¹⁴C,⁹⁰Sr/⁹⁰Y,²³⁹Pu) can usually be measured only by indirect methods. However, some beta emitters, especially those with high energy emissions such as ³²P or ⁹⁰Sr/⁹⁰Y, can sometimes be measured

‘directly’ via the bremsstrahlung produced. Such bremsstrahlung measurements, because of their relatively high minimum detectable activities (see Appendix II), are not usually employed for routine monitoring.

3.21. Direct measurements, where they are possible, offer the advantage of a rapid and convenient estimate of the total activity in the body or a defined part of the body at the time of measurement; when it is sufficiently sensitive, for example for ¹³¹I and

137Cs, direct measurement of body or organ content is therefore to be preferred.

Whole body and individual organ measurements are less dependent on biokinetic models than indirect monitoring measurements, but they suffer from greater calibration uncertainties, especially for low energy photon emitters. Direct measurements may necessitate the worker being removed from work involving radiation exposure for the period over which the retention characteristics are measured, and usually need special, well shielded (and therefore expensive) facilities and equipment.

3.22. Direct measurements are useful in qualitative as well as quantitative determi- nations of radionuclides in a mixture that may have been inhaled, ingested or injected.

In addition, direct measurements can assist in identifying the mode of intake by determining the distribution of activity in the body [13, 14]. Sequential measurements, where they are possible, can reveal the redistribution of activity and give information about the total body retention and the biokinetic behaviour of radionuclides in the body.

3.23. Indirect measurements generally interfere less with worker assignments, but require access to a radiochemical analytical laboratory; such a laboratory may also be used for measuring environmental samples, but high level (e.g. reactor water chem- istry) and low level (e.g. bioassay or environmental samples) measurements should be performed in separate laboratories. Excreta measurements determine the rate of loss of radioactive materials from the body by a particular route, and must be related to the body content and intake by a biokinetic model. Because of the ability of radiochemical analyses to detect low levels of activity, measurements of excreta usually give sensitive detection of activity in the body.

3.24. Measurements of air samples can be difficult to interpret, because they measure the concentration of radionuclides in the air at the location of the sampler, not necessarily in the breathing zone of the worker. However, a personal air sampler (PAS) placed on the worker’s lapel or protective headgear can collect a sample that is

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representative of the activity concentration in air which the worker has inhaled, except in cases where the sample comprises only a few particles. Air concentration measurements, combined with assumptions about breathing rates and volumes and measured exposure times, can be used to estimate the intake. However, the use of PASs allows only estimates of intake and cannot be used to refine a dose estimate based on individual retention characteristics. Furthermore, PAS measurements cannot be repeated if an analytical result is suspect or is lost. They can, however, provide estimates of intakes for radionuclides such as ¹⁴C (in particulate form),²³⁹Pu,²³²Th and

235U, for which direct and other indirect methods of assessment of body activity are not sufficiently sensitive. This method of monitoring depends for its interpretation on the dose coefficients and the derived air concentrations (DACs), which are defined in Section 2 and discussed further in Section 7. Dose coefficients and DACs for various chemical forms of selected radionuclides are provided in Tables A–I and A–II.

3.25. Particle size influences the deposition of inhaled particulates in the respiratory tract and so information about the distribution of particle sizes is needed for the correct interpretation of bioassay results and subsequent dose assessment. In many situations the airborne particle size distribution should be determined using cascade impactors or other methods. As a minimum, air sample measurements should include measurement of the concentration of the respirable fraction of airborne particulates.

Some models for interpreting PAS results discriminate against non-respirable particles [15]. In general, the more site and material specific information that is available, the better will be the dose assessment.

3.26. Measurement methods have limits of detection arising from the presence of naturally occurring radioactive materials, from statistical fluctuations in counting rates, and from factors related to sample preparation and analysis. Appendix II describes the concepts of minimum significant activity (MSA) and minimum detectable activity (MDA), which are used to characterize the limits of detection of any measurement method.

Frequency of monitoring

3.27. As stated in para. I.35 (Appendix I) of the BSS [2]: “The nature, frequency and precision of individual monitoring shall be determined with consideration of the magnitude and possible fluctuations of exposure levels and the likelihood and magnitude of potential exposures.” In order to determine the appropriate frequency and type of individual monitoring, the workplace should be characterized. The radionuclides in use and, if possible, their chemical and physical forms should also be known. If these forms are likely to change under accident conditions (e.g. the release of uranium hexafluoride into the atmosphere results in the production of HF and uranyl fluoride),

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this should also be considered. The chemical and physical forms (e.g. particle size) of the material determine its behaviour on intake and its subsequent biokinetics in the human body. These in turn determine the excretion routes and rates, and hence the type of excreta samples to be collected and their frequency.

3.28. A consideration in setting a bioassay sampling schedule is to minimize the uncertainty in estimates of intake due to the unknown time of an intake within the monitoring period. The ICRP [8, 9] recommend that monitoring periods should generally be selected so that assuming an intake to have occurred at the mid-point of the monitoring period would not lead to underestimation of the intake by a factor of more than three.

3.29. Another consideration in scheduling a worker for monitoring, whether by direct or indirect methods, is to ensure that an intake above a predetermined level is not

‘missed’ [16]. An intake could be missed if, as a result of radioactive decay and biological clearance, the body content or daily excretion of the radionuclide were to decline to a level below the minimum significant activity (MSA) of the measurement during the time interval between the intake and the measurement (see Appendix II for further details). The fraction of an intake remaining in the body for direct measurement or being excreted from the body for indirect measurement,m(t), depends on both the physical half-life and the biokinetics of the radionuclide, and is a function of the time since intake. Thus, an intake Iand the resulting committed effective dose E(50) would be missed if I×m(t) is less than the MSA. Typically, the frequency of monitoring should be set so that intakes corresponding to more than 5% of the annual dose limit are not missed.

3.30. The frequency of monitoring will thus be driven to a great extent by the sensitivity of the measurement technique. Although techniques for measurement should be as sensitive as possible, the costs of using the most sensitive techniques and the short- est possible sampling interval should be balanced against the radiation detriment associated with doses that might be underestimated or missed if less sensitive methods or less frequent measurements are used.

3.31. In any case, the bioassay method and measurement frequency adopted should be capable of detecting an intake that results in a specified fraction of the dose limit.

Sometimes this goal cannot be realized because of a lack of analytical sensitivity, unacceptably long counting times for direct measurements, or unacceptably short sampling intervals for excreta collection, particularly in the case of faecal sampling to monitor inhalations of insoluble particulates. In such cases, additional methods such as improved workplace monitoring and personal air sampling should be used to ensure adequate worker protection.

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Reference levels

3.32. Reference levels are helpful in the management of operations. They may be expressed in terms of measured quantities or in terms of other quantities to which measured quantities can be related, and if they are exceeded, some specified action or decision should be taken. The various types of reference level are described in the related Safety Guide [3]. In relation to intakes of radionuclides, reference levels are generally based on the committed effective dose E(50). The appropriate fraction of the dose limit corresponding to each type of reference level (see below) should be established with other sources of exposure taken into account. Investigation levels and recording levels are of relevance to monitoring for internal contamination in the case of occupational exposures.

Investigation level

3.33. An investigation level is “the value of a quantity such as effective dose, intake or contamination per unit area or volume at or above which an investigation should be conducted” [2]. For intakes of radionuclides, the investigation level relates to a value of committed effective dose above which a monitoring result is regarded as sufficiently important to justify further investigation. The investigation level set by management will depend upon the objectives of the programme and the type of investigation to be carried out.

3.34. For routine monitoring, the investigation level for an intake of a radionuclide is set in relation to the type and frequency of monitoring, as well as the expected level and variability of intakes. The numerical value of the investigation level depends on a knowledge of the conditions in the workplace. An investigation level may be set for individuals involved in a particular operation, either routinely or on an occasional basis, or may be devised for individuals within a workplace without reference to a particular operation.

3.35. As an example, for a routine operation with routine monitoring, an investigation level IL may be set on the basis of a committed effective dose of 5 mSv (0.005 Sv) from a year’s intakes. Thus, for N monitoring periods per year, the investigation level (in Bq) for the intake of any radionuclide j in any monitoring period would be given by:

where e(g)_jis the appropriate dose coefficient for inhalation or ingestion.

IL_j

N e(g)j

= 0 005.

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Recording level

3.36. A recording level is defined as “a level of dose, exposure or intake specified by the regulatory authority at or above which values of dose, exposure or intake received by workers are to be entered in their individual exposure records” [2]. As an example, the recording level RL for an intake of a radionuclide could be set to correspond to a committed effective dose of 1 mSv (0.001 Sv) from a year’s intakes. Thus, for N monitoring periods per year, the recording level for intake of radionuclide jin a monitoring period would be given by:

Derived levels

3.37. The quantities actually measured in individual bioassay programmes are radionuclide activities in the body or excreta samples, and it is therefore convenient to establish reference levels for the measurement results themselves. These are termed derived investigation levels (DILs) and derived recording levels (DRLs). They are measurement results that imply radionuclide intakes or committed effective doses at the corresponding reference levels. Derived investigation and recording levels are calculated separately for each radionuclide. They are specific to the radiochemical form in the workplace, and are a function of time since intake. For the examples given above,

where t₀, the typical time elapsed since intake when a bioassay sample is taken, is usually calculated as 365/2Ndays, based on the assumption that the intake occurs at the mid-point of the monitoring period, and

Even if the resulting dose is below that associated with the recording level, the measurement results should always be maintained in the radiation monitoring records for the workplace and for the individual [17] (see also Section 8). In cases of worker exposure to external radiation or to multiple radionuclides, management may decide to reduce the derived levels for individual radionuclides appropriately.

DRL_j

N e(g)j m t

= 0 001 ×

0

. ( )

DIL_j

N e(g)j m t

= 0 005 ×

0

. ( )

RL_j

N e(g)j

= 0 001.

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Use of material specific and individual specific data

3.38. Biokinetic models for most radionuclides in their commonly encountered forms, with reference parameter values, have been published by the ICRP (see Section 6). These models are based on Reference Man [18] and the observed behaviour of radionuclides in humans and animals. They have been developed for defined chemical forms of radionuclides and are generally of use for planning purposes. As mentioned above, the particular workplace conditions should be characterized to determine which forms are actually present. It is likely that, in some circumstances, the chemical or physical forms of the radionuclides in use in a given workplace will not correspond to the reference biokinetic models. In this circum- stance, material specific models may need to be developed.

3.39. If intakes are small, for example corresponding to a few per cent of the dose limit, the reference models are likely to be adequate for estimating the resulting doses. However, if the estimate of an intake corresponds to about a quarter or more of the dose limit, biokinetic model parameters specific to the material(s) and individual(s) in question may need to be developed to estimate the committed effective dose more accurately. Such biokinetic models can be developed from sequential direct and indirect measurements of the exposed workers. Analysis of workplace air and surface contamination samples can also assist in the interpretation of bioassay measurements, for example by measuring the ratio of ²⁴¹Am to ^239+240Pu when direct measurement of ²⁴¹Am in the lung is used to assess plutonium intakes or for assessing the solubility of inhaled particles [13, 14].

3.40. A common example of the need for material specific information is where the size of the particles that a worker would be likely to inhale differs significantly from the assumption of 5 µm AMAD recommended by the ICRP as a default value for the workplace [19]. In this case, the fractions of inhaled radioactive materials deposited in the various regions of the respiratory tract would have to be determined from the ICRP respiratory tract model (see Section 6) [19] and an appropriate dose coefficient calculated. More specific information may also be needed on the solubility characteristics of the material after inhalation or ingestion as appropriate. This can be obtained from experimental studies in animals or by in vitro solubility studies.

Retrospective determination of particle characteristics following an exposure may be difficult and consideration should be given to obtaining material specific information when setting up worker monitoring programmes.

3.41. Even if all of the assumptions in the reference biokinetic models are appropriate for a given workplace, there will still be differences between individuals in excretion rates and other biokinetic parameters for the same intake of a radionuclide. The

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variability between individuals, and even in the daily excretion rate for the same individual, will often be more significant than the differences between a reference biokinetic model and one developed specifically for a given individual. To reduce some of this variability, collection periods for excreta samples should be sufficiently long, for example 24 hours for urine and 72 hours for faeces. The use of individual specific model parameters should be rare under routine circumstances.

Task related monitoring

3.42. Task related monitoring is, by definition, not routine, i.e. it is not regularly sched- uled. Such monitoring is conducted to provide information about a particular operation and to give, if necessary, a basis for decisions on the conduct of the operation. It is particularly useful when short term procedures are carried out under conditions which would be unsatisfactory for long term use. Task related monitoring is usually conducted in the same way as routine monitoring, unless the circumstances of the operation dictate otherwise, for example if the radionuclides involved may be different or if the probability or potential magnitude of internal exposure may be significantly greater.

Special monitoring

3.43. Special monitoring may be necessary as a result of a known or suspected exposure, or an unusual incident, such as a loss of containment of radioactive materials as indicated by an air or surface sample, or following an accident. It is most often prompted by a result of a routine bioassay measurement that exceeds the derived investigation level. It may also result from occasional samples such as nose blows, swipes or other monitoring.

3.44. Special monitoring prompted by an incident is not usually conducted any differently from a routine measurement in terms of measurement techniques, although improved sensitivity or a faster processing time may be needed. The laboratory should be advised that the sample analysis or the direct measurement has priority over routine measurements, and the frequency of subsequent monitoring may be changed. The laboratory should also be informed that samples may have a higher than normal level of activity, so that the measurement technique can be tailored to the special monitoring situation and any necessary precautions taken to prevent contamination of other samples.

ASSESSMENT FOLLOWING ACCIDENTS OR INCIDENTS

3.45. There will be situations involving the use of radioactive material in which the operational controls break down. Accidents or incidents may result in releases of

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radioactive materials into the working environment with the potential for high doses to the workforce.

3.46. After an accident has occurred, the radiological consequences may be compli- cated by trauma or other health effects incurred by the workers. Medical treatment of injuries, especially those that are potentially life threatening, generally takes priority over radiological operations, including exposure assessment. In such cases, post-accident exposure assessment should be conducted when the situation has been brought under control.

3.47. Once assessment of internal exposure has commenced, as much information should be gathered as is practicable. For example, information will be needed on the time and nature of the incident and the radionuclides involved, and on the timing of bioassay samples and measurements of body activity. This information may be necessary not only for exposure assessment, but also to assist in medical assessment, to guide medical treatment of the victim (which may include chelation therapy or wound excision), and to assist later in reconstruction of the accident or incident itself and in long term medical follow-up of the victim [20, 21].

3.48. Because intakes associated with accidents or incidents can result in committed effective doses which approach or exceed dose limits, individual and material specific data are normally needed for exposure assessment. These data include information on the chemical and physical forms of the radionuclide(s), the particle size, airborne concentrations, surface contamination levels, the retention characteristics in the individual affected, nose blows, face wipes and other skin contamination levels and external dosimetry results. The various items of data will often seem to be incon- sistent or contradictory, particularly if the intake period is uncertain. An adequate assessment of dose can be made only after considering all of the data, resolving the sources of inconsistency as far as is possible, and determining the most likely and worst possible scenarios for the exposure and the magnitude of any intake.

Direct and indirect methods

3.49. The primary factor in deciding between direct and indirect methods of internal exposure monitoring after an accident or an incident will be the radiological characteristics of the radionuclides involved. If the victim is externally contaminated with gamma emitting radionuclides, direct measurements should normally be delayed until the victim has been decontaminated, both to prevent interference with the measurement and also to avoid contamination of the direct measurement facility [22, 23]. Occasionally, the urgency of assessment may preclude complete decontamination, in which case the individual could be wrapped in a clean sheet to minimize

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contamination of the facility. The result of this initial direct measurement would set an upper limit for the body content, but more measurements would be needed after further decontamination [24]. External contamination with alpha or pure beta emitters will normally not interfere with direct measurements, unless bremsstrahlung is produced by the beta emitter(s). External contamination will not interfere with indirect methods, provided that care is taken to avoid transfer of contamination to excreta samples. On rare occasions, intakes may be so high that special techniques are needed for either direct or indirect measurements to avoid interference with equipment response, such as excessive electronic dead times [22, 23].

3.50. Following an accident or incident, analyses of samples of urine and faeces should be considered to verify the intake of radioactive material. However, the results of such analyses are frequently difficult to interpret, because of the potential for multiple routes of intake and imprecise knowledge of the amount of radionuclide transferred to the blood from points of intake. Excreta sample measurements are generally not useful for intake assessment immediately after an accident or incident because of the delay between intake and excretion; this is particularly the case for faecal excretion. In addition, rapid early components of urinary excretion can be difficult to interpret as they are not fully defined in some biokinetic models. Nevertheless, all excreta should be collected following an accident or incident; the early detection of radioactive materials in a urine sample can be a useful indication of the solubility of the radioactive material involved and of the potential for effective treatment.

Excreta analyses can be the only reliable method of assessing intakes if large amounts of external contamination interfere with direct measurements.

3.51. In view of the general principle of emphasizing non-invasive procedures, invasive procedures such as blood sampling will usually be justified only in accident situations in which large intakes may have occurred. Blood sampling can provide data on the solubility and biokinetics of the material involved, but is generally of limited value for providing quantitative estimates of the intake because of the rapid clearance of most radionuclides to other tissues.

3.52. Workplace monitoring samples, such as air filters and surface contamination wipes, should be analysed to determine the radionuclides involved, isotopic ratios and their physicochemical characteristics.

Follow-up monitoring

3.53. Both direct and indirect follow-up monitoring programmes should be conducted at reasonable intervals for an extended period after an accident or incident.

This information will help in establishing the biological half-lives of radionuclides in

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the body tissues and their excretion rates. This, in turn, can help to improve the accu- racy of dose assessment.

Schedule of sampling

3.54. Following an accident or incident, excreta samples for indirect monitoring should be collected until such time as a reasonable estimate can be made of the temporal pattern of excretion. If decorporation therapy, such as the administration of chelating agents, is used [20], samples should continue to be collected in order to determine the effectiveness of the treatment. Once excretion patterns have stabilized, individual samples collected during the course of a day may be combined into 24-hour samples, and appropriate aliquots taken for analysis.

3.55. If direct measurements are feasible, they should be continued at regular intervals if the subject’s medical condition permits. The frequency of direct measurements will be determined by the clearance and decay rates of the internally deposited radioactive materials. Sequential direct measurements of specific organs or body regions can also assist in determining the biokinetics of the activity. For example, sequential measurements of inhaled ²⁴¹Am can demonstrate the clearance from lung and translocation to bone and liver [25]. In the case of deposits in cuts or wounds of some insoluble forms of radioactive materials, follow-up monitoring may reveal deposition in regional lymph nodes as a consequence of lymphatic clearance, with slow clearance from these sites [26, 27].

4. DIRECT METHODS

INTRODUCTION

4.1. The most accurate assessments of internal dose can be made when the distribution and total body content of an incorporated radionuclide can be determined reli- ably by direct in vivo counting of emissions from the body. Nevertheless, biokinetic modelling of retention and biophysical modelling of energy deposition may still be needed to calculate the intake and the committed effective dose, so direct methods can also depend on the interpretation of rates of excretion, which often vary markedly over time and between individuals.

4.2. Direct measurement is possible when the incorporated radionuclide(s) emit(s) penetrating radiation (normally X ray or gamma photons, including bremsstrahlung)

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of sufficient energy and yield to be detectable (Appendix II) outside the body. A detailed description of the methods commonly used in direct measurement can be found in Ref. [7]. For most in vivo counting applications, photon detectors are positioned at specified locations around the body, usually with at least partial shielding of the detector and/or the subject to reduce interference from ambient external sources.

Arc geometry Chair geometry

Standing geometry (scanning or static)

Scanning bed geometry Stretcher geometry

FIG. 1. Various geometries used for whole body monitoring.

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MEASUREMENT GEOMETRIES

4.3. A variety of physical arrangements of detectors has been developed to serve specific purposes. For radionuclides which are distributed throughout the body, counting of the whole body, or a large fraction of it, provides the greatest sensitivity.

Whole body counting is carried out either using a static geometry, with one or more detectors, or by scanning — moving the subject with respect to static detectors or moving detectors around a static subject. Static geometries commonly comprise an array of detectors distributed along a standing or supine subject, or a single detector directed towards the centre of a subject on a tilted chair or curved frame. Some examples of counting geometries are shown in Fig. 1.

4.4. For other radionuclides which are at least temporarily concentrated in particular organs or tissues of the body, monitoring of specific sites is recommended.

Examples are radioiodine, which is taken up by the thyroid, and inhaled radioactive particles which are retained in the lungs. Localized monitoring is also recommended when intake is through a wound, or when there are other reasons for determining the distribution of the radionuclide(s) within the body.

4.5. In all cases the method should be to compare the signal measured from the subject with that obtained under the same conditions from an anthropomorphic phantom, or other surrogate, containing known quantities of the radionuclide in question. The distribution of the radionuclide in the calibration phantom should match that expected in the human subject as far as possible, although some measurement techniques are more sensitive than others to this distribution. Whole body counting is unlikely to fail completely to detect a significant amount of localized activity, but might not provide an accurate estimate of the amount or give good information on its spatial distribution.

METHODS OF DETECTION

4.6. A variety of detection systems are in use for different purposes. Inorganic crys- tals of high atomic number materials, usually thallium-activated sodium iodide, NaI(Tl), are commonly used to detect energetic photons (above 100 keV), such as those emitted by many fission and activation products. Scintillations produced by the crystal’s interaction with high energy photons are detected by photomultiplier tubes;

these generate electronic pulses which are processed to produce a spectrum reflecting that of the radiation absorbed by the crystal. This type of measurement system is most suited to cases where a small number of radionuclides are present; the energy resolu- tion is limited, so that even deconvolution techniques may be unable to determine the radionuclides giving rise to a complex spectrum, such as that from a fresh fission

Assessment of Occupational Exposure Due to Intakes of Radionuclides | IAEA

IAEA SAFETY STANDARDS

SERIES

Assessment of

Occupational Exposure Due to Intakes of

Radionuclides

SAFETY GUIDE

No. RS-G-1.2

IAEA SAFETY RELATED PUBLICATIONS

ASSESSMENT OF OCCUPATIONAL EXPOSURE

DUE TO INTAKES OF

RADIONUCLIDES

ASSESSMENT OF

OCCUPATIONAL EXPOSURE DUE TO INTAKES OF

RADIONUCLIDES

SAFETY GUIDE

JOINTLY SPONSORED BY THE

INTERNATIONAL ATOMIC ENERGY AGENCY AND THE INTERNATIONAL LABOUR OFFICE

SAFETY STANDARDS SERIES No. RS-G-1.2

FOREWORD by Mohamed ElBaradei

Director General

PREFACE

CONTENTS

1. INTRODUCTION

2. DOSIMETRIC QUANTITIES

3. MONITORING PROGRAMME

4. DIRECT METHODS